Nuclear fuel having improved fission product retention properties

ABSTRACT

A nuclear fuel based on UO 2 , ThO 2  and/or PuO 2  having improved retention properties for fission products. The fuel comprises a metal such as Cr or Mo able to trap oxygen in order to form an oxide having a free formation enthalpy equal to or below that of the superstoichiometric oxide or oxides (U, Th)O 2+x  and/or (U, Pu)O 2+x  (O&lt;x≦0.01). Thus, it is possible to trap oxygen atoms released during the fission of U, Th and/or Pu. This leads to an increase in the retention level of the fission products and a possibility of obtaining a high burn-up of nuclear fuel elements.

FIELD OF THE INVENTION

The present invention relates to nuclear fuels based on UO, ThO₂ and/orPuO having improved fission product retention properties.

BACKGROUND OF THE INVENTION

In the use of nuclear fuels based on oxide, particularly uranium oxide,one of the problems caused is due to the release of fission gases in thefuel element during the operation of the reactor, because these fissionproducts must be kept in the fuel element, particularly in the actualfuel pellets, so as to limit the internal pressure of the sheaths orcans and the interaction of the fission products with the latter.

Therefore, at present, the burn-up of nuclear elements is limited to 50GWj/t of U in order not to exceed the threshold beyond which the releaseof fission gases becomes significant.

However, operators of electronuclear reactors, particularly pressurizedwater reactors (PWR), wish to optimize control of the nuclear fuel byincreasing burn-up of the uranium dioxide pellets contained in the rodsin order to achieve minimum values of 60 to 70 GWj/tU.

Research carried out up to now for obtaining such an improvement hasused procedures for increasing the size of the uranium dioxide grains,because it has been found that the gas quantity released by anirradiated large grain fuel is less than that released by an irradiatedsmall grain fuel. Use has also been made of procedures for formingprecipitates in the nuclear fuel in order to anchor the fission gases onsuch precipitates.

In order to obtain an increase in the size of the uranium dioxidegrains, it is possible to add additives such as iO, NbO, CrO, AlO, VOand MgO to the uranium dioxide powder subject to fritting or sinteringin order to activate its crystal growth, provided that the sinteringtakes place under a wet hydrogen atmosphere so that the added oxidequantity remains in solution in the uranium dioxide and is not reducedto a metallic element. The use of such additives for obtaining a largegrain microstructure is, e.g., described by Killeen in Journal ofNuclear Materials, 88, 1980, pp 177-184, Sawbridge et al in Report CEGBRD/B/N 4866, July 1980 and Radford et al in Scientific Paper81-7D2-PTFOR-P2, 1981. However, the use of certain additives of thistype can lead to an increase in the diffusion coefficients of cationsand fission gases in the uranium dioxide, which is unfavorable for theretention of the fission products and does not make it possible to takefull advantage of the large grain microstructure.

Another procedure for improving the retention rate of nuclear fuelsconsists of dispersing in the uranium dioxide grains nanoprecipitates ofa second phase for ensuring the anchoring of the fission products onsuch second phase. Nanoprecipitates of this type can consist ofmagnesium oxide inclusions, as described by Sawbridge et al. in Journalof Nuclear Materials, 95, 1980, pp. 119-128, and in FR-A-2 026 251.

SUMMARY OF THE INVENTION

The present invention makes use of a method different from thatdescribed hereinbefore for improving the retention rate of fissionproducts in a nuclear fuel. This method consists of trapping the oxygenatoms released by the fission of the uranium and/or plutonium atoms, soas to maintain the O/U (Th, Pu) or O/M stoichiometry with M=U+Pu or U+Thor U+Pu+Th of the fuel at 2 and thus prevent a rise in the diffusioncoefficients in the fuel and a reduction of its thermal conductivity.

Thus, the increase in the diffusion coefficients is a mechanism leadingto the accumulation of fission products at the grain boundaries,followed by release of these fission products. In the same way, areduction of the thermal conductivity of the fuel is prejudicial,because it has the effect of increasing the temperature of the fuel forthe same linear power and consequently both reducing the solubility ofthe fission products and favoring their diffusion.

The invention also relates to a process for improving the retention offission products within a ceramic nuclear fuel based on UO₂, ThO₂ and/orPuO₂, which consists of including in the ceramic nuclear fuel at leastone metal able to trap oxygen by forming an oxide having a freeformation enthalpy at the operating temperature T of the nuclear reactorbelow the free formation enthalpy at the same temperature T of thesuperstoichiometric oxide or oxides of formulas (U, Th)O_(2+x) and/or(U, Pu)O_(2+x), in which x is such that 0<x≦0.01.

The use of such a metallic additive consequently makes it possible tomaintain the O/U (Th or Pu) or O/M ratio defined hereinbefore of thenuclear fuel at a value of 2 and in this way to avoid an increase in thediffusion coefficients, which remain at a low value, and a reduction inthe thermal conductivity of the fuel. Thus, a high fission productretention rate is obtained.

This procedure can be combined with known methods of increasing the sizeof the UO₂ and/or PuO₂ and/or ThO₂ grains and forming precipitates foranchoring the fission gases, which is very interesting and makes itpossible to improve the performance characteristics of the fuel.

In the case of uranium dioxide-based fuels, the free formation enthalpyof the superstoichiometric oxide UO_(2+x) with 0<x≦0.01 can be expressedin oxygen potential and calculated on the basis of the law of Lindemerand Besmann, as described in Journal of Nuclear Materials, 130, 1985,pp. 473-488. In this case, and as is indicated on p. 480 thereof, theoxygen potential ΔG(O₂) of the above-defined superstoichiometric oxidecan be expressed in J/mole according to the following formula:

    -360 000+214 T+4 RTLn[2x(1-2x)/(1-4x).sup.2 ]

in which R is the molar constant of the gases, T is the temperature inKelvins and x is as defined hereinbefore.

Moreover, for uranium dioxide-based fuels, the metal included in thefuel must be able to form an oxide having an oxygen potential defined bythe formula: ΔG(O₂)=RT Ln (pO₂) in which R is the molar constant of thegases, T the reactor operating temperature and p(O₂) the partial oxygenpressure, equal to or below the above-estimated value for UO_(2+x) inaccordance with the Lindemer and Besmann law. Examples of suitablemetals are Cr, Mo, Ti, Nb and U.

The invention also relates to a fuel for nuclear reactors comprising aceramic material based on UO₂, ThO₂ and/or PuO₂ in which is dispersed atleast one metal able to trap oxygen and having the characteristics givenhereinbefore.

According to the invention, the oxide-based ceramic material can beconstituted by UO₂, ThO₂ PuO₂ or mixtures thereof, the mixed oxide UO₂--PuO₂ or UO₂ --ThO₂, mixed oxides based on UO₂ and other oxides such asoxides of rare earths or mixed oxides based on PuO₂.

Preferably, the ceramic material is based on UO₂ and the dispersed metalis able to form an oxide having an oxygen potential below the oxygenpotential of UO_(2+x), as described hereinbefore.

In general, to obtain a burn-up of 60 GWj/t⁻¹, the dispersed metalrepresents 0.1 to 2% by weight of the fuel material. Preferably, themetal is chromium and represents 0.1 to 1 or better still 0.2 to 0.5% byweight of the fuel material.

Moreover, the fuel material can also comprise additives such as TiO₂,Nb₂ O₅, Cr₂ O₃, Al₂ O₃, V₂ O₅ and MgO, in order to increase the size ofthe fuel grains and/or aid the anchoring of the fission products, aswell as other additives, e.g., SiO₂, in order to improve otherproperties.

The fuel material according to the invention can be prepared byconventional sintering or fritting processes by adding to the ceramicmaterial powder to be sintered the metal, either in metallic form, or inthe form of an oxide or oxygenated compound.

In the first case, after shaping the powder by cold compression,sintering is carried out in a dry hydrogen atmosphere, e.g., having awater content below 0.05 volume % so as not to oxidize the metal.

In the second case, if it is wished to simultaneously obtain a sizeincrease of the UO₂, ThO₂ and/or PuO₂ grains, use is made of an oxide oroxygenated compound quantity which may or may not exceed the solubilitylimit of the oxide or oxygenated compound in UO₂, ThO₂ and/or PuO₂ atthe sintering temperature. After shaping the powder by cold compression,sintering takes place under wet or humidified hydrogen, e.g., having awater content above 1 volume %, in order to preserve the oxide duringsintering and activate crystal growth. After sintering, the sinteredmaterial undergoes a reduction treatment under dry hydrogen, e.g.,having a water content below 0.05 volume %, in order to reduce the oxideor oxygenated compound to metal.

In the second case, this manner of operating makes it possible to obtaina large grain microstructure (diameter>40 μm) with:

either intragranular, nanometric metallic precipitates (diameter<100 nm)if the added oxide or oxygenated compound quantity is below thesolubility limit,

or intragranular, nanometric metallic precipitates (diameter<100 nm) andmicrometric, metallic precipitates (diameter>0.3 μm) if the added oxideor oxygenated compound quantity exceeds the solubility limit.

However, if in the second case it is not wished to simultaneously obtaina large grain structure following the shaping of the powder by coldcompression, sintering takes place under a dry hydrogen atmosphere,e.g., having a water content below 0.05 volume %, in order tosimultaneously reduce the oxide or oxygenated compound to metal. In thiscase, intragranular, micrometric, metallic precipitates are obtained(diameter>0.3 μm).

In all these cases, the shaping of the powder by cold compression, e.g.,in order to form pellets, can be carried out in a conventional manner byuniaxial compression, e.g., under pressures of 200 to 700 MPa.

A temperature of 1600 to 1750° C. is normally used for sintering.

When there is a supplementary reduction heat treatment, the latter canbe carried out at temperatures of 1300 to 1750° C.

The ceramic material powder including the metallic additive in the formof an oxide or oxygenated compound can be prepared by mixing powders ofthe constituents or by atomization-drying processes using a slipcontaining the additive in the form of salt in solution, or bycoprecipitation of a uranium salt and a salt of the additive.

Therefore the process of the invention makes it possible to takeadvantage not only of the oxygen trapping capacity of the added metal,but also of the properties of the metallic oxides in order to activatethe UO₂ crystal growth and improve the retention of fission productswithin the fuel.

BRIEF DESCRIPTION OF THE DRAWINGS

Other features and advantages of the invention can be gathered from thefollowing description and with reference to the attached drawings.

FIGS. 1, 2 & 3 are graphs illustrating the evolution of the oxidepotentials of various oxygenated compounds (in kJ/mole) as a function ofthe temperature (in °C).

FIG. 4 is a micrograph of a fuel material according to the invention.

FIG. 5 is a micrograph illustrating oxygen trapping in a fuel materialaccording to the invention.

FIG. 6 is a micrograph given for comparison purposes in order to showthe structure of a prior art fuel material following managed oxidation.

FIG. 7 is a micrograph of a fuel material according to the inventionhaving a small grain structure.

FIG. 8 is a micrograph of a fuel material according to the inventionhaving a large grain microstructure.

DESCRIPTION OF PREFERRED EMBODIMENT

FIG. 1 shows the oxygen potential in kJ/mole calculated on the basis ofthe Lindemer and Besmann formula for UO₂, as well as forsuperstoichiometric oxides UO_(2+x) and substoichiometric oxidesUO_(2-x), as a function of the temperature in °C.

FIG. 2 shows the evolution of the oxygen potential (in kJ/mole) for theCr/Cr₂ O₃ pair as a function of the temperature (in °C.), and it can beseen that, throughout the temperature range in question, the oxygenpotential of the oxide is below that of the superstoichiometric oxidesUO_(2+x) of FIG. 1.

FIG. 3 shows the evolution of the oxygen potential (in kJ/mole) for MoO₂as a function of the temperature, and it can be seen that it is stillbelow that of the superstoichiometric oxides UO_(2+x) at the sametemperatures.

Consequently these two elements are suitable as a metal able to trapoxygen for fuel materials based on UO₂ and the following examplesillustrate the use of the two elements with UO₂. In all the examples,use is made of a UO₂ powder with an average grain size of 0.5 to 100 μm.

EXAMPLE 1

In this example preparation takes place of UO₂ pellets incorporatingmicrometric, metallic precipitates of Cr.

100 g of UO₂ powder are mixed together with 0.1 g of metallic Cr powderhaving an average grain size below 2 μm and then the mixture is broughtinto pellet form by uniaxial compression at 350 MPa, the matrix beinglubricated in a hydraulic press. The pellets are then placed in amolybdenum boat and sintered at 1700° C. for 4 h under dry hydrogen.This gives a small UO₂ grain microstructure with micrometric, metallicprecipitates of Cr.

FIG. 4 is a micrograph illustrating this structure with a 600Xmagnification. It is clearly possible to see the intergranular orintragranular metallic precipitates (white particles), and the electrondiffraction pattern confirms the metallic character of these inclusions.

In order to verify the behavior of said fuel for trapping oxygen, amanaged oxidation takes place of the pellets by heat treatment at 700°C. in a helium atmosphere having 0.01 vol % oxygen, under conditionsmaking it possible to achieve in the case of pure oxide an average O/Uratio of 2.024.

FIG. 5 is a micrograph with a 400× magnification illustrating thestructure of the fuel material having undergone the oxidation. It can beseen that the fuel material has trapped the oxygen and has no phasesother than the previously obtained UO₂ matrix.

For comparison purposes, FIG. 6 shows the micrograph of a uraniumdioxide pellet obtained under the same conditions as in example 1, butwithout any chromium addition and when it has undergone the same managedoxidation for obtaining the average O/U ratio of 2.024. FIG. 6 showsthat there are U₄ O₉ needles in the UO₂ matrix.

Thus, by comparing FIGS. 5 and 6, it is possible to see theeffectiveness of the metallic chromium inclusions, which have preventedthe transformation of UO₂ into U₄ O₉.

EXAMPLE 2

In this example preparation takes place of uranium dioxide nuclear fuelpellets having a small UO₂ grain microstructure with micrometric,metallic Cr precipitates.

In this case, 100 g of UO₂ powder are mixed with 0.15 g of Cr₂ O₃ powder(with a grain size below 2 μm), followed by the formation of pelletsfrom the mixture and they are sintered as in Example 1, under a dryhydrogen atmosphere.

In this case, the added chromium oxide is reduced to metallic chromiumduring the sintering under dry hydrogen and has not activated thecrystal growth of UO₂ in order to form a large grain microstructure.Thus, a small grain microstructure is obtained with metallic Crprecipitates. FIG. 7 shows this structure.

EXAMPLE 3

In this example, preparation takes place of a nuclear fuel having a UO₂small grain microstructure with metallic Cr precipitates.

Preparation takes place of a powder by the atomization-drying of a slipcontaining 150 g of UO₂, 0.6 g of a soluble chromium salt: (NH₄)₂ CrO₄and 250 g of distilled water. The powder obtained is then calcined for 2h in an alumina boat at 400° C. in an alumina laboratory tubular furnaceunder an argon flow (300 ml/min) in order to transform the chromium saltinto Cr₂ O₃. This is followed by the shaping of the powder andsintering, as in Example 1, under a dry hydrogen atmosphere.

In this case, the oxygenated compound of the chromium is reduced duringsintering into metallic chromium, so that it cannot serve as anactivator for UO₂ crystal growth. Thus, a UO₂ small grain microstructureis obtained with metallic chromium precipitates.

EXAMPLE 4

In this example, preparation takes place of a nuclear fuel having a UO₂large grain microstructure with nanometric, micrometric, metallicprecipitates of Cr.

A powder is prepared by atomization-drying, as in Example 3, using 1.5 gof (NH₄)₂ CrO₄, i.e., a Cr₂ O₃ content above the Cr₂ O₃ solubility limitin UO₂ at 1700° C. The powder obtained is treated in accordance withExample 3, being calcined for 2 h in an alumina boat at 400° C. in analumina laboratory tube furnace under an argon flow (300 ml/min). It isthen brought into the form of pellets by uniaxial compression at 350MPa, as in Example 1. Sintering then takes place under a hydrogenatmosphere humidified with 1.7 vol. % water, at 1700° C. and for 4 h inorder to keep the chromium in oxide form and assist the increase in theUO₂ grain size.

After sintering, an annealing treatment takes place at 1300° C. for 5 hand under dry hydrogen having a water content below 0.05 vol. % in orderto reduce the Cr₂ O₃ oxide to metallic chromium.

Maintaining the Cr₂ O₃ in oxide form during sintering has made itpossible to use it as an activator for crystal growth and in this way toobtain a large grain microstructure and the annealing treatment underdry hydrogen has then made it possible to reduce Cr₂ O₃ to metallicchromium and consequently obtain nanometric, micrometric, metallicprecipitates.

The microstructure of the material obtained under these conditions isillustrated in FIG. 8, where it is possible to see the large grains 1 ofUO₂ and the micrometric chromium inclusions 5. The nanometric chromiuminclusions are revealed by electron diffraction.

EXAMPLE 5

A powder is prepared as in Example 3 by atomization-drying, but using0.2 g of (NH₄)₂ CrO₄, i.e. a Cr₂ O₃ equivalent content below thesolubility limit of Cr₂ O₃ in UO₂ at 1700° C. This is followed by thecompression of the powder in the form of pellets and sintering as inExample 4 to obtain a large grain microstructure due to the maintainingof the chromium in oxide form. This is followed by an annealingtreatment as in Example 4 for reducing Cr₂ O₃ into metallic chromium.

In this case, a large grain UO₂ microstructure is obtained withnanometric metallic precipitates of Cr, because there was no Cr₂ O₃excess for forming metallic, micrometric precipitates during thereduction.

EXAMPLE 6

This example adopts the same operating procedure as in Example 4, butuse is made of 1.5 g of (NH₄)₂ CrO₄ and 0.04 g of ultrafine SiO₂ in slipcontaining 150 g of UO₂ and 250 g of distilled water. The powderobtained by atomization-drying is compressed in pellet form and thensintered in a humidified hydrogen atmosphere and subjected to anannealing treatment under dry hydrogen, under the same conditions as inExample 4. This gives a large grain UO₂ microstructure with metallicchromium precipitates and a silica phase at the grain boundaries.

EXAMPLE 7

In this example a mixture of 100 g of UO₂ and 0.6 g of MoO₃ is preparedby cogrinding in a metallic uranium ball jar, followed by thecompression of the powder mixture to pellet form and sintering under thesame conditions as in Example 1.

In this case, the molybdenum oxide is reduced to molybdenum duringsintering and it is not possible to active the crystal growth of the UO₂grains. Thus, a small grain UO₂ microstructure is obtained withmicrometric, metallic precipitates of Mo.

EXAMPLE 8

A powder is obtained by atomization-drying of an aqueous suspensionconstituted by 150 g of UO₂ and 7.7 g of ammonium heptamolybdate (NH₄)₆Mo₇ O₂₄, 4H₂ O and 250 g of distilled water. The powder is then treatedas in Example 1. This gives a small grain UO₂ microstructure withmicrometric, metallic Mo precipitates.

I claim:
 1. A fuel material for a nuclear reactor, said fuel materialbeing devoid of ceric oxide and comprising a ceramic material based onstoichiometric UO₂ Th₂ and/or PuO₂ in which is dispersed 0.1 to 2% byweight of at least one elemental metal able to trap oxygen for formingan oxide, whose free formation enthalpy is equal to or below the freeformation enthalpy of the superstoichiometric oxide UO_(2+x), (U,Th)O_(2+x) and/or (U, Pu)O_(2+x) with x such that 0<x≦0.01 at thetemperature reached in the nuclear reactor.
 2. The fuel materialaccording to claim 1, wherein the metal is able to form an oxide havingan oxygen potential defined by the formula:

    ΔG(O.sub.2)=RTLn (pO.sub.2)

in which R is the molar constant of the gases, T is the nuclear reactortemperature in Kelvins and pO₂ is the partial oxygen pressure equal toor below

    360 000+214 T+4 RTLn {2x(1-2x)/(1-4x).sup.2 }

in which R and T have the meanings given hereinbefore and x is such that0<x≦0.01.
 3. The fuel material according to claim 1 or 2, wherein themetal is chosen from the group consisting of Cr, Mo, Ti, Nb and U. 4.The fuel material according to claim 3, wherein the metal is Cr and theCr content of the fuel material is 0.1 to 1% by weight.
 5. The fuelmaterial according to claim 4, wherein the chromium content is 0.2 to0.5% by weight.
 6. The fuel material according to claim 1 or 2, whereinthe ceramic material is a mixed oxide of UO₂ and a rare earth oxide.